Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
SSINS No.: 6835
IN 85-59
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
July 17, 1985
Information Notice No. 85-59: VALVE STEM CORROSION FAILURES
Addressees: All nuclear power reactor facilities holding an operating
license (OL) or a construction permit (CP).
Purpose:
This information notice is provided to alert recipients of a potentially
significant problem pertaining to stress corrosion failures of valve stems
and shafts; items that are not routinely examined. It is suggested that
recipients review the information for applicability to their facilities and
consider actions, if appropriate, to preclude a similar problem occurring at
their facilities. However, suggestions contained in this information notice
do not constitute NRC requirements; therefore, no specific action or written
response is required.
Description of Circumstances:
There have been four instances where cracks were found in 410 stainless
steel valve stems. These instances involved different licensees and
different manufacturers. Such cracks cannot be observed without the
disassembly of the valves, and the valve operability test programs do not
provide a means of early detection. In three of these instances, the cracks
grew until the stem sheared when the valve was activated. Such failures can
prevent the system from performing its safety function.
Uncontrollable leakage from the stem packing of several Velan globe valves
was reported by Oconee 1 in December 1971. Disassembly and examination
revealed cracks for the entire length of the stems and more than half of the
diameter in depth. In order to prevent cracking, 410 stainless steel needs
to be tempered immediately after hardening, but several batches were not
tempered. Ultimately, 2600 stems were replaced in the 1-1/2-, 1- and
1/2-inch valves, using 17-4PH and 300-series stainless steel materials.
A 20 inch Anchor/Darling gate valve stem snapped while being manually opened
at Brunswick 2 on August 4, 1982. There was pitting of the 410 stainless
steel in the gland packing section and the crack had initiated from one of
these pits. The cross-section area of the stem of the suppression pool
suction valve had been reduced by 70% by intergranular stress corrosion
cracking (IGSCC). The material had a higher hardness than specified as a
result of improper heat treatment.
8507120320
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IN 85-59
July 17, 1985
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Excessive hardness is associated with cracking and corrosion. The
manufacturer replaced five lots of valve stems with proper heat-treated 410
stainless steel, and there have not been any further problems.
An injection valve in the low-pressure coolant injection (LPCI) system broke
in two places during disassembly at Browns Ferry 3 on February 28, 1984. One
break was below the stem packing area and the other was at the gate
connection. Over 50% of the cross-section of the stem of these 24-inch
Walworth valves had been lost in these areas by IGSCC. The stem had higher
hardness than specified. New stems made from 17-4 PH stainless steel were
installed.
Linear indications were discovered on three main steam isolation valve
(MSIV) shafts at Farley 1 on February 29, 1984. The indications were from 1
to 13 inches, long, contained thick oxides, and were located in the packing
gland area. The MSIVs were Atwood-Morrill 32-inch swing check valves and the
shaft hardness exceeded specifications. 17-4 PH stainless steel also was
used as the replacement material for the shafts.
Although 410 stainless steel is defined as a stainless steel because of its
alloy content, it is really a high chromium, very hardenable steel. Cooling
this material in air from the 1700-to-1900F temperature range results
in a surface hardness of up to HRC 45 and high internal stresses. Tempering
the hard and brittle martensite produces a softer and more ductile
composition that has much less chromium available for intergranular
corrosion resistance. Tempering in the 700-to-1050F range is not
recommended because it results in low and erratic impact properties and poor
resistance to corrosion and stress corrosion.
The following conclusions were reached:
1. The actual hardness of the 410 stainless steel valve stems and shafts
was higher than specified and higher than documented.
2. The excessive hardness is associated with intergranular stress
corrosion cracking.
3. The cracking occurred in internal areas where there could be
concentrations of corroding chemicals, such as at the gland packing.
4. The oxides found in the cracks showed that the cracks occurred during
service and grew slowly.
5. The cracks were not detected by the routine valve operability test
programs, but were only discovered by actual failures or after
disassembly during refueling outages.
6. Failure of these valves would make the specific safety system
inoperable.
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IN 85-59
July 17, 1985
Page 3 of 3
No specific action or written response is required by this information
notice. If you have any questions about this matter, please contact the
Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: P. Cortland, IE
(301) 492-4175
Attachment: List of Recently Issued IE Information Notices
